TPTF, JAERI (SET facility)
JapanMain characteristics
Number of simulated fuel rods:25
Heat method:Indirect
Experimental conditions
Test details:
Technical subjects : (1) Heat transfer in uncovered core a. Heat transfer in low flow rate region b. Steam cooling with moisture and flow c. Natural convection heat transfer by stagnant steam (2) Two phase flow in horizontal pipe a. Slug and stratified flow in a horizontal hot leg. (3) Heat transfer in SG a. Steam condensation and counter current flow limiting (CCFL) in SG primary side b. Liquid level effect on heat transfer in SG secondary side c. Effect of location auxillary feedwater (FM) inlet (4) calibration of two-phase flow instruments to be used in LSTF
Pressure min:3 MPa
Pressure max:12.7 MPa
Temperature max:330 °C
Facility geometry
Keywords General:
- Non Nuclear Core
Keywords Core:
- Rod Simulator
Keywords Steam Separator:
- Separator
- Heat exchanger (HX)
Keywords Others:
- Valve
- Boiler
- Pump(Simulator)
Test section details:
Two phase flow intruments of Horizontal pipe test section Instruments are located in the 2m pipe sections which are connected by Gray Locks to form a 10m horizontal pipe test section. A pitot tube and a correlation flow meter are used to measure the liquid and steam velocities. Five point CPT(conductivity probe with thermocouple) rakes and three-beam r-densitometers will be used to measure the liquid levels and the fluid density.
References
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Nuclear Engineering and Design 96, 1986.
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Experiment Data of ROSA-III Integral Test Run 7341; Single Failure Series Test No.4; Full ECCS, JAERI-M 83-043 ,1983
http://jolissrch-inter.tokai-sc.jaea.go.jp/search/servlet/search?3019945&language=1
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Developmental Assessment of RELAP5/MOD3 Code against ROSA-IV/TPTF Horizontal Two-Phase Experiments, Kukita et al., JAERl-M 90-053, March 1990.
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H. Nakamura et al.: "System Description for ROSA-IV Two Phase Flow Test Facility" (TPTF), JAERl-M 83-042, March 1983
https://jopss.jaea.go.jp/search/servlet/search?3019944&language=1
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H. Kumamaru et al.: "Investigation of Uncovered Bundle Heat Transfer under High Pressure Boil-off Conditions" ; Nuclear Engineering and Design 96 (1986) 81-94, September 1985
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H.Nakamura et al., Int. J. Muhiphase flow 13 (2) pp. 145.159, Phase and velocity distributions and holdup in high-pressure steam/water stratified flow in a large diameter horizontal pipe, 1987
https://www.sciencedirect.com/science/article/pii/0301932287900267
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H.Kumamaru, Y.Koizumi, K.Tasaka, Nucl. Engg. Design 102 pp.71- 84 , Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle, 1987
https://www.sciencedirect.com/science/article/pii/0029549387902688
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Y.Koizumi, H.Kumamaru, T.Yonomoto, K.Tasaka, Nucl. Engg. Design 99 pp. 157-165, Post-dryout heat transfer of high-pressure steam-water two-phase flow, 1987
https://www.sciencedirect.com/science/article/pii/0029549387901178
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Y. Koizumi, T. Yonomoto, H. Kumamaru, K. Tasaka, J. Nucl. Sci. Technol.25 pp. 104- 106, Post-Dryout Heat Transfer Coefficient of High-Pressure Steam-Water Two-Phase Flow in Multi-Rod Bundle, 1988
https://www.tandfonline.com/doi/abs/10.1080/18811248.1988.9733561
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T.Yonomoto, Y. Koizumi, K.Tasaka, J. Nucl. Sci. Technol.24 pp. 798-810, Onset Criterion for Liquid Entrainment in Reflooding Phase of LOCA, 1987
https://www.tandfonline.com/doi/abs/10.1080/18811248.1987.9735883
Phenomena
| Relevance | Phenomenon |
|---|---|
| Entrainment and Deentainment | |
| Global Multidimensional Fluid Temperature, Void and Flow Distribution | |
| Heat Transfer : Core | |
| Heat Transfer : SG | |
| Phase separation | |
| Quench Phenomena | |
| Stratification in horizontal pipes |